1. Field of the Invention
The present invention relates generally to fuel assemblies for a nuclear reactor and, more particularly, is concerned with a boiling water reactor (BWR) fuel assembly having a water cross with coolant flow entry ports which minimize uncertainties in the amount of flow entering the water cross.
2. Description of the Prior Art
Typically, large amounts of energy are released through nuclear fission in a nuclear reactor with the energy being dissipated as heat in the elongated fuel elements or rods of the reactor. The heat is commonly removed by passing a coolant in heat exchange relation to the fuel rods so that the heat can be extracted from the coolant to perform useful work.
In nuclear reactors generally, a plurality of the fuel rods are grouped together to form a fuel assembly. A number of such fuel assemblies are typically arranged in a matrix to form a nuclear reactor core capable of a self-sustained, nuclear fission reaction. The core is submerged in a flowing liquid, such as light water, that serves as the coolant for removing heat from the fuel rods and as a neutron moderator. Specifically, in a BWR the fuel assemblies are typically grouped in clusters of four with one control rod associated with each four assemblies. The control rod is insertable in between the fuel assemblies for controlling the reactivity of the core. Each such cluster of four fuel assemblies surrounding a control rod is commonly referred to as a fuel cell of the reactor core.
A typical BWR fuel assembly in the cluster is ordinarily formed by an N-by-N array or bundle of the elongated fuel rods. In the present disclosure, an exemplary array of 64 fuel rods in an 8.times.8 configuration is shown. The fuel rods are supported in laterally spaced-apart relation and are encircled by a tubular outer channel member having a generally rectangular cross-section. Examples of such fuel assemblies are illustrated and described in U.S. Pat. Nos. 3,689,358 to Smith et al and 3,802,995 to Fritz et al, and in Canadian Patent No. 1,150,423 to Anderson et al, as well as in the patent applications cross-referenced above.
In a fuel assembly of this type, the fuel rods in the central region of the bundle thereof may be undermoderated and overenriched. In order to remedy this condition by increasing the moderation in this region of the assembly, several water cross arrangements have been proposed. See, for example, the above cross-referenced Barry et al, Doshi, Lui and Taleyarkhan patent applications.
As disclosed in the aforementioned cross-referenced applications, the central water cross is located within the outer channel member and is formed of four radial panels which together form a cruciform water flow channel which divides the fuel assembly into four separate elongated compartments, with the bundle of fuel rods being divided into mini-bundles disposed in the respective compartments. The water cross thus provides a centrally disposed cross-shaped path for the flow of subcooled neutron moderator water within the channel along the length of, but separate from, adjacent fuel rods in the mini-bundles thereof.
The fuel rods of each mini-bundle extend in laterally spaced-apart relationship between an upper tie plate and a lower tie plate. The fuel rods, selectively connected with the upper and lower tie plates in a known manner, comprise a separate fuel rod subassembly within each of the compartments of the outer channel member. A plurality of grids axially spaced along the fuel rods of each fuel rod subassembly maintain the fuel rods in their laterally spaced relationships. In many applications, the water cross has approximately the same axial length as the fuel rod subassemblies, extending between the upper and lower tie plates thereof.
As mentioned initially, coolant is passed along the fuel rods for removing heat therefrom. In the design of BWRs in the United States, subcooled water enters a bottom nozzle of the fuel assembly through a side entrance. Thereafter, the water is distributed upwardly into the four mini-bundles and the water cross. Typically, flow through the water cross is approximately 9 to 10 percent of the total flow through the fuel bundle. Due to the side entry characteristics of these BWRs, it has been found that significant maldistribution of flow can result at the inlet to each fuel assembly. This, along with complex nozzle flow patterns and crud buildup effects, results in uncertainties in the mass flow actually entering the water cross.
Maldistribution affects the operation of the individual fuel rod mini-bundles. As an example, a deficiency in coolant inlet flow to one mini-bundle can lead to increased boiling of the coolant/moderator in that mini-bundle with a consequent deterioration in the moderation efficiency of the coolant and the loss of neutron fissioning efficiency of that mini-bundle. That loss of power production must be compensated in one of the other mini-bundles in the assembly. This means that that mini-bundle must in turn operate closer to the fuel safety limits than desired. If the water cross has irregular or uncertain flow, the operation of the entire fuel rod assembly may be drastically affected. If the coolant flow to the water cross is too low, boiling of that coolant/moderator can result. This reduces the moderation efficiency of the coolant and the fissioning efficiency of the fuel assembly. That loss of power production leads to the under-utilization of that fuel assembly with its economic penalties. It also means that the remaining fuel assemblies in the core must make up the loss of power production. Therefore, those assemblies are forced to operate closer to the fuel safety limits than desired. Conversely, if the water cross flow is too large, this means that a greater fraction of the coolant flow than desired is diverted to the water cross which is not available to provide heat transfer cooling of the fuel rods in that assembly.
A plot of the water cross inlet orificing loss coefficient vs. orificing flow area is set forth in FIG. 6. In current designs the fuel bundle operates in the relatively steep portion of the curve at S.sub.1. Thus, a small change in orifice area may cause a large increase in flow loss, which would in turn lead to possible boiling of water cross moderator fluid and attendant effects as noted above. Further, as the total flow of the bundle is decreased (at constant power) the water cross flow is even more dramatically affected, so that the undesirable results become more likely. If the orificing to the water cross is controlled so as to operate in the relatively flap part S.sub.2 of the curve of FIG. 6, changes in orificing losses would not create significant changes in the flow to the water cross.
Consequently, the need exists for further improvement of the BWR fuel assembly so as to eliminate or significantly minimize water cross flow uncertainty, mini-bundle inlet flow maldistribution and crud buildup sensitivity, and thereby avoid the undesirable effects which accompany these conditions.